Filtered generated 401 hits.
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2011:05 Overview and Evaluation of the NESC Projects for Fracture Assessments of Nuclear Components
The Network for Evaluation of Structural Components (NESC) was started in 1993 and since then six different projects have been carried out. A seventh NESC-project is still in progress. The overall objective of these projects has been to study the reliability of the entire process of structural integrity assessment within an international framework. The network is coordinated by the European...
Content type: Publications -
2010:22 Research and Development Program in Reactor Diagnostics and Monitoring with Neutron Noise Methods, Stage 16
This report constitutes Stage 16 of a long-term research and development program concerning the development of diagnostics and monitoring methods for nuclear reactors. Results up to Stage 15 were reported in SKI and SSM reports, as listed below and in the Summary. The results have also been published in international journals and have been included in both licentiate- and doctor’s degrees.
Content type: Publications -
2011:04 Evaluation of the Technical Basis for New Proposals of Fatigue Design of Nuclear Components
During the recent years fatigue analysis procedures for nuclear components have been investigated. The most common method so far has been the American code ASME III. The basis for the current design procedures in ASME III is quite old and has now been evaluated against modern data leading to the proposal of modified design curves. Also the effect of the environment has been the subject of...
Content type: Publications -
2010:45 Evaluation of models for cyclic plastic deformation – A literature study
Analysis of nuclear components subjected to cyclic plastic deformation is normally based on the American standard ASME III. With the focus to improve the modeling of the stress-strain response under non-monotonic loading, several models for the analysis of cyclic plastic deformation have been developed in recent years. However, experience shows that the analysis and evaluation of plastic...
Content type: Publications -
00:48 The Use of Risk Based Methods for Establishing ISI-Priorities for Piping Components at Oskarshamn 1 Nuclear Power Station
Björn Brickstad SKI English...
Content type: Publications -
99:55 Risk Based Test Interval and Maintenance Optimisation Application and Uses
Erik Sparre SKI English...
Content type: Publications -
2022:04 Warm Pre-Stressing mechanisms
SSM perspective Background Swedish nuclear power plants (NPP) were originally designed and analyzed with an assumed operating period of approximately 40 years. As Swedish NPPs reach their original technical design life the safety of the additional operating time has to be verified. With increased operating time, the shift in the fracture toughness curve of the reactor pressure vessel (RPV)...
Content type: Publications -
2022:09 Assessment of modelling approaches for axial gas flow inside LWR fuel rods
SSM perspective Background The fuel rod analysis program SCANAIR has been developed by IRSN (Institut de Radioprotection et de Sûreté Nucléaire) for analysis of reactivity initiated accidents (RIA) in light water reactors. The Swedish Radiation Safety Authority (SSM) has access to SCANAIR in exchange for annual contributions for its further development. The development and administration...
Content type: Publications -
2017:19 Numerical simulations of headed anchors break in reinforced and non-reinforced concrete structures, Phase 2
Background In the design of anchorage equipment in concrete structures the beneficial effects of reinforcement are, in most cases, not considered. The American ASCI 349-06 code opens up for a more detailed analysis where the beneficial impact of reinforcement on anchor capacity can be taken into account. How this analysis can be done is, however, not explicitly described in the code. The...
Content type: Publications -
2022:14 Weld residual stress and strain measurements on a mock-up with single layer strip cladding common in reactor pressure vessels
Summary Residual stress and strain measurements were performed on a mock-up consisting of a 160 mm thick low-alloy ferritic steel plate of type SA-508 Class 3 with a single layer weld deposited strip cladding of austenitic stainless steel E308L. The cladding was manufactured with submerged arc strip welding adapted to a procedure used for reactor pressure vessels in the 1970s. Comprehensive...
Content type: Publications